Radiation defect clustering in graphite studied by synchrotron radiation scattering

The formation of radiation defect clusters as interstitial dislocation loops in pyrographite irradiated by 20-MeV carbon ions in a cyclotron has been studied by means of synchrotron radiation scattering. The experimental data are interpreted in terms of the Krivoglaz model and the model of deformation domain formation by dislocation loops. Parameters of the pyrographite crystal lattice in the deformed state upon ion irradiation have been determined using experimental data on the shift and broadening of the Bragg peaks. The results of experiments and their theoretical analysis indicate that the phenomena of the crystal lattice densification (shrinkage) and the radiation-induced swelling (increase in volume) observed for graphite irradiated by fast neutrons in a nuclear reactor can be studied in model experiments on charged particle accelerators using irradiation of the samples with carbon ions at much lower doses and significantly shorter exposure times.     

Dislocation density-based constitutive model for the mechanical behaviour of irradiated Cu

Performance degradation of structural steels in nuclear environments results from the formation of a high number density of nanometre-scale defects. The defects observed in copper-based alloys are composed of vacancy clusters in the form of stacking fault tetrahedra and/or prismatic dislocation loops that impede the motion of dislocations. The mechanical behaviour of irradiated copper alloys exhibits increased yield strength, decreased total strain to failure and decreased work hardening as compared to their unirradiated behaviour. Above certain critical defect concentrations (neutron doses), the mechanical behaviour exhibits distinct upper yield points. In this paper, we describe the formulation of an internal state variable model for the mechanical behaviour of such materials subject to these (irradiation) environments. This model has been developed within a multiscale materials-modelling framework, in which molecular dynamics simulations of dislocation–radiation defect interactions inform the final coarse-grained continuum model. The plasticity model includes mechanisms for dislocation density growth and multiplication and for irradiation defect density evolution with dislocation interaction. The general behaviour of the constitutive (homogeneous material point) model shows that as the defect density increases, the initial yield point increases and the initial strain hardening decreases. The final coarse-grained model is implemented into a finite element framework and used to simulate the behaviour of tensile specimens with varying levels of irradiation-induced material damage. The simulation results compare favourably with the experimentally observed mechanical behaviour of irradiated materials.     

Kinetics of dislocation ensembles in deformable irradiated materials

The development of plastic instability in the initial deformation stages of irradiated materials is studied. The dependence of the fraction of dislocations which overcome obstacles in the dynamic regime (dislocation “channeling”) on the degree of radiation hardening (irradiation dose) and the dislocation velocity is analyzed. It is shown that this effect plays a role in radiation embrittlement of reactor materials. Fiz. Tverd. Tela (St. Petersburg) 40, 1631–1634 (September 1998)     

晶体塑性有限元方法研究辐照对多晶铜力学性能的影响

将辐照硬化理论与晶体塑性理论结合, 运用ABAQUS有限元分析软件模拟辐照后多晶铜的拉伸过程。分析辐照效应对材料屈服强度、硬化过程、晶体变形等力学性能的影响, 研究位错密度的演化及空间分布规律。数值模拟表明: 辐照效应提高多晶铜的屈服应力, 影响不同阶段的硬化和软化现象; 辐照剂量增大导致位错密度增殖总体变缓, 空间不均匀度增大; 晶体的塑性变形及晶体转动也受到辐照的影响, 在相同的应变条件下, 辐照剂量越大, 晶体塑性变形程度越小, 塑性变形分布不均匀度变大, 同时晶体转动程度及转动角离散度增大。     

强流脉冲电子束作用下金属锆的微观结构与应力状态

利用强流脉冲电子束 (HCPEB) 技术对金属纯锆进行表面处理, 采用X射线衍射, 扫描电子显微镜及透射电子显微镜详细分析了辐照诱发的表层微观结构和缺陷. X射线分析结果表明, HCPEB辐照后在材料表层诱发幅值为GPa量级的压应力, 并形成{0002}, {1012}, {1120}及{1013}织构. 表层微观结构观察表明, 与其他金属材料不同, HCPEB辐照在材料表层诱发的熔坑数量极少, 多次轰击甚至几乎没有表面熔坑的形成. 此外, 在快速的加热和冷却状态下, 在表面熔化层形成大量的超细晶粒结构, 同时诱发马氏体相变和强烈的塑性变形. 1次HCPEB辐照后表层内形成的变形微结构以位错为主, 孪晶数量较少; 5 次辐照样品的位错密度迅速增高, 孪晶数量也显著增加; 10次辐照后样品中的变形微结构以变形孪晶为主, 且出现二次孪晶现象. 表层晶粒内部变形的晶体学特征不仅决定了表层的织构演化行为, 而且还起到细化晶粒的作用, 为纯锆及锆合金表面强化提供了一条有效的途径. 关键词： 强流脉冲电子束 纯锆 微观结构 应力状态     

质子辐照对纯铝薄膜微观结构的影响

利用透射电子显微镜(TEM)详细分析了不同剂量的质子辐照纯铝薄膜样品的微观结构, 质子的能量E=160 keV.实验表明,质子辐照能够在Al薄膜中诱发空位位错圈,在实验范围内,位错密度随辐照剂量的增加而增加;质子辐照在1×10¹¹—4×10¹¹/mm²范围内随辐照剂量的增加,位错圈数量密度以及位错圈尺寸都随之增加.在较高剂量6×10¹¹/mm²辐照下,位错圈数量密度减小,但其尺寸显著 关键词： 质子辐照 空位簇缺陷 位错圈 微观结构     

In situ and tomographic observations of defect free channel formation in ion irradiated stainless steels

The effects of heavy-ion irradiation on dislocation processes in stainless steels were investigated using in situ irradiation and deformation in the transmission electron microscope as well as post mortem electron tomography to retrieve information on the three-dimensional dislocation state. Irradiation-induced defects were found to pose a strong collective barrier to dislocation motion, leading to dislocation pileups forming in grain interiors and at grain boundaries. The passage of multiple dislocations along the same slip plane removes the irradiation defects and leads to the eventual formation of a defect-free channel. These channels are composed of densely tangled dislocation networks which line the channel-matrix walls as well as residual dislocation debris in the channel interiors. The structures of the dislocation tangles were found to be similar to those encountered in later stages of deformation in unirradiated materials, with the exception that they developed earlier in the deformation process and were confined to the defect free channels. Also, defect free channels were found to widen through both source widening as well as complex cross-slip mechanisms.     

强流脉冲电子束诱发纯钛表面的微观结构及应力状态

利用强流脉冲电子束(HCPEB)装置对金属纯钛进行轰击,采用X射线衍射,扫描电子显微镜及透射电子显微镜技术详细分析了轰击样品表层的结构和缺陷. X射线衍射分析表明, HCPEB能够在材料表层诱发幅值为 GPa量级的压应力,并在(100), (102)和(103)晶面出现择优取向.表层微观结构的观察表明: HCPEB轰击后材料表层发生了马氏体相变,形成了大量的片状马氏体组织; 此外, HCPEB轰击还在辐照表面诱发了强烈的塑性变形,一次轰击后,晶粒内部的塑性变形以(100)晶面的位错滑移为主,位错密度显著提高;多次轰击后,样品变形结构发生变化,变形孪晶的数量明显增多. 这些变形微结构不仅影响表层的织构演化行为,而且还能细化晶粒,进而提高材料表面硬度, 为HCPEB技术进行纯钛表面强化提供了一条有效的途径.     

强流脉冲电子束作用下金属纯Cu的微观结构状态——变形结构

为了研究金属的超快变形机理,利用强流脉冲电子束(HCPEB)技术对多晶纯Cu进行了辐照处理,并利用透射电子显微镜对HCPEB诱发的表面微结构进行了表征.实验结果表明,HCPEB轰击多晶纯Cu后,在轰击表层诱发了幅值极大的应力和极高的应变速率.1次HCPEB轰击材料表层的变形结构以交滑移形成的位错胞和位错缠结结构为主;多次轰击后平行的位错墙和孪晶是该区域的主要变形结构特征;原子面的扩散乃至位错攀移可在晶界和孪晶界上形成台阶结构.根据各自区域的变形结构特征,对相应的变形机理进行了探讨. 关键词： 强流脉冲电子束 多晶Cu 变形结构 孪晶     

Effect of gamma irradiation on the polymer electrolyte PEO-NH4ClO4

Irradiation affects the structures of materials at different scales, thus changing physical and chemical properties. We study here the effect of gamma irradiation at different doses on the polymer electrolyte PEO-NH₄ClO₄. Optical micrographs show cracks in the irradiated samples and impedance spectroscopy measurements indicate reduced ion-conductivity at room temperature but slight enhancement at higher temperature. At high frequencies, the real part of the admittance shows a power-law variation; the exponent, which is a measure of self-similarity of the structure, is reduced on irradiation. The overall results point to a more disordered structure at higher radiation doses.     

Crystal plasticity-based constitutive modelling of irradiated bcc structures

A constitutive crystal plasticity model is proposed and developed for the inelastic deformation of irradiated bcc ferritic/martensitic steels. Defects found in these irradiated materials are used as substructure variables in the model. Insights from lower length- and time-scale simulations are used to frame the kinematic and substructure evolution relations of the governing deformation mechanisms. Models for evolution of mobile and immobile dislocations, as well as interstitial loops (formed due to irradiation), are developed. A rate theory-based approach is used to model the evolution of point defects generated during irradiation. The model is used to simulate the quasi-static tensile and creep response of a martensitic steel over a range of loading histories.     

氦、氘对纯铁辐照缺陷的影响

在核聚变堆的辐照环境中, 核嬗变产物氢、氦对结构材料的抗辐照性能将产生很大的影响. 本实验采用离子注入和电子辐照模拟研究了氦和氘对具有体心立方结构的纯铁的影响. 采用离子加速器在室温分别对纯铁注入氦离子和氘离子, 经500℃时效1 h后在高压电镜下进行电子辐照.结果表明: 室温注氦和室温注氘的纯铁在500℃时效后分别形成间隙型位错环和空位型位错环. 在电子辐照下, 间隙型位错环吸收间隙原子而不断长大, 而空位型位错环吸收间隙原子不断缩小. 通过计算位错环的变化速率发现, 空位型位错环比间隙型位错环吸收了更多的间隙原子, 即室温注氘纯铁的位错偏压比室温注氦纯铁的偏压参量大, 这意味着相同实验条件下空位型位错环对辐照肿胀的贡献大于间隙型位错环对辐照肿胀的贡献. 利用氦-空位复合体和氘-空位复合体的结构, 分析了注氦和注氘后在纯铁中形成不同类型位错环的原因. 关键词： 氦 氘 辐照损伤 位错环     

Analysis of the parameters of a submicron dislocation structure in metals subjected to severe plastic deformation

Equations of dislocation kinetics are used to quantitatively compare the mechanisms of formation and evolution (with deformation) of cellular dislocation structures at moderate strains and of submicron block dislocation structures at high plastic strains. In both cases, the formation of nonuniform dislocation structures is a result of dislocation self-organization, more specifically, the self-organization of statistically random dislocations during the formation of cellular structures and the self-organization of geometrically necessary dislocations (which appear due to the nonuniform character of plastic deformation on the micron scale) during the formation of block structures.     

Microstructure and hardening effect of pure tungsten and ZrO2 strengthened tungsten under carbon ion irradiation at 700℃

Microstructure evolution and hardening effect of pure tungsten and W-1.5%ZrO₂ alloy under carbon ion irradiation are investigated by using transmission electron microscopy and nano-indentation. Carbon ion irradiation is performed at 700 ℃ with irradiation damages ranging from 0.25 dpa to 2.0 dpa. The results show that the irradiation defect clusters are mainly in the form of dislocation loop. The size and density of dislocation loops increase with irradiation damages intensifying. The W-1.5%ZrO₂ alloy has a smaller dislocation loop size than that of pure tungsten. It is proposed that the phase boundaries have the ability to absorb and annihilate defects and the addition of ZrO₂ phase improves the sink strength for irradiation defects. It is confirmed that the W-1.5%ZrO₂ alloy shows a smaller change in hardness than the pure tungsten after being irradiated. From the above results, we conclude that the addition of ZrO₂ into tungsten can significantly reduce the accumulation of irradiated defects and improve the irradiation resistance behaviors of the tungsten materials.     

超临界水冷堆燃料包壳材料的辐照损伤研究进展

超临界水冷堆（SCWR）是第四代核电站的主力堆型之一,高温、高压、超临界水环境下的辐照损伤问题是其燃料包壳材料面临的最大挑战。SCWR燃料包壳候选材料主要包括锆合金、奥氏体不锈钢、铁素体/马氏体不锈钢、镍基合金、ODS合金五大类,奥氏体不锈钢是最有希望的候选材料。介绍了近年来在这个领域国际上的主要研究进展。作者所在团队也对多种SCWR的候选材料进行了辐照损伤研究,包括:镍基合金C-276和718、铁素体/马氏体钢P92、奥氏体不锈钢AL-6XN和HR3C。对AL-6XN的氢离子辐照实验发现,辐照产生的缺陷主要是间隙型位错环,伯格斯矢量为1/3<111>,在较高剂量（5～7 dpa）辐照下,出现空洞肿胀。在氢滞留的影响下,位错环有着独特的演化规律,总结提出了位错环的四阶段演化过程。The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.     

Irradiation hardening of reactor pressure vessel steels due to the dislocation loop evolution

The irradiation hardening of reactor pressure vessel steels due to the formation of dislocation loops is analyzed. The analysis is based on the original model for the nucleation and subsequent evolution of dislocation loops in irradiated materials. The loop formation in displacement cascades is taken into account, along with the homogeneous clustering of point defects. The loop evolution is shown to contribute mainly to the athermal component of the yield stress, which is determined by interaction of gliding dislocations with strong barriers. Irradiation-induced hardening is evaluated as a function of irradiation dose and temperature, dose rate, material parameters and initial microstructure. The model results are compared with experimental data for neutron irradiated pressure vessel steels of various grades and with empirical low power expressions of the yield stress increase with increasing irradiation dose.     

γ辐照后硼硅酸盐玻璃的吸收光谱研究

玻璃固化体是用来固化放射性废物的硅酸盐。作为放射性废物处理的第一道工程屏障,它的耐辐照性能尤其引人注目。本工作使用硼硅酸盐玻璃模拟玻璃固化体材料,用不同剂量的γ射线辐照硼硅酸盐玻璃模拟天然放射性对固化体的辐照损伤。通过测量和分析辐照后硼硅酸盐玻璃的吸收光谱,证实了辐照后在玻璃中产生了E'缺陷,非桥氧空位色心、过氧自由基以及过桥氧联接等缺陷。此外,还得出了不同微观结构随吸收剂量的演化关系。对于辐照后产生的缺陷,它们的浓度都随吸收剂量的增加而增大。同时,发现玻璃在辐照后其吸收光谱的带隙随着剂量的增加而逐渐变窄;而当吸收剂量大于等于105 Gy时,玻璃的带隙则达到饱和值。Vitrification is one kind of silicates which is used for immobilization of high-level waste (HLW). As the first engineered barrier of HLW disposition, its anti-irradiation characteristic is particularly noticeable. Vitrification is replaced by borosilicate glass to investigate radiation effect, and the irradiation damage generated by natural radioactivity in vitrification is simulated by different doses gamma rays on borosilicate glass. By measuring and analyzing the absorption spectrum of irradiated borosilicate glasses, it is confirmed that E'defect, non-bridging oxygen hole center, peroxy dangling bond and bridge oxygen link, etc. are induced in borosilicate glass after irradiation. Furthermore, the relations between the defects and absorbed doses are shown. For the concentrations of these defects induced by irradiation, all of them increase with absorbed dose. Meanwhile, absorption band gap in borosilicate glass after irradiation decreases with absorbed dose and the band gap becomes saturated when absorbed dose is equal to or greater than 105 Gy.     

Radiation-induced transmission loss in optical materials at infrared wavelengths

Information concerning the effect of irradiation on the optical properties of materials in the infrared, sub-mm-and mm-wavelength regions is of increasing importance in fusion plasma diagnostics. The radiation induced transmission loss of a number of materials has therefore been investigated at wavelengths in the ranges 200 nm to 40 μm and 0.23 to 2.0 mm. The samples were irradiated with doses of up to 10¹⁰ rad in a nuclear reactor. While germanium shows considerable transmission loss at doses as low as 10⁶ rad, and the transmission of TPX decreases at 10⁹ rad, other materials, e.g. fused quartz and possibly ZnSe, exhibit sufficient radiation hardness for use in fusion plasma diagnostics.     

Molecular dynamics simulation of the interaction between a mixed dislocation and a stacking fault tetrahedron

A high number-density of nanometer-sized stacking fault tetrahedra are commonly found during irradiation of low stacking fault energy metals. The stacking fault tetrahedra act as obstacles to dislocation motion leading to increased yield strength and decreased ductility. Thus, an improved understanding of the interaction between gliding dislocations and stacking fault tetrahedra are critical to reliably predict the mechanical properties of irradiated materials. Many studies have investigated the interaction of a screw or edge dislocation with a stacking fault tetrahedron (SFT). However, atomistic studies of a mixed dislocation interaction with an SFT are not available, even though mixed dislocations are the most common. In this paper, molecular dynamics simulation results of the interaction between a mixed dislocation and an SFT in face-centered cubic copper are presented. The interaction results in shearing, partial absorption, destabilization or simple bypass of the SFT, depending on the interaction geometry. However, the SFT was not completely annihilated, absorbed or collapsed during a single interaction with a mixed dislocation. These observations, combined with simulation results of edge or screw dislocations, suggest that defect-free channel formation in irradiated copper is not likely by a single dislocation sweeping or destruction process, but rather by a complex mix of multiple shearing, partial absorption and defect cluster transportation that ultimately reduces the size of stacking fault tetrahedra within a localized region.     

Impact of gamma radiation on the eruption rate of rat incisors

The present work aims to test the effect of gamma radiation on the rate of eruption of rat incisors. One hundred and five adult male albino rats were used and irradiated at different gamma doses. The effects of irradiation were investigated by numerical measurements of eruption rate, histological investigation using light microscope and spectral analysis using Fourier Transform Infra-Red (FTIR). No detectable changes were observed in the groups with smaller radiation doses. There was a significant decrease in the eruption rate starting from the 4?Gy radiation dose. The observation of histological sections revealed disturbance in cellular elements responsible for eruption as well as periodontal disturbance in the samples irradiated with 4 and 6?Gy. FTIR Spectroscopy of control group and the group irradiated by 0.5?Gy showed similar absorption bands with minor differences. However, samples irradiated by 1?Gy showed significant changes in both molecular structure and conformation related to carbonates and hydroxyl groups. From the previous results, it could be concluded that gamma irradiation negatively affects the eruption rate of the rat incisors especially with higher doses.